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31	Innovative Nuclear Energy Systems	ABS-76
PERFORMANCE ANALYSIS ON THORIUM FUEL BASED SMALL MODULAR REACTOR (SMR) Amila Amatullah (a*), Sidik Permana (b,c), Dwi Irwanto (b,c), Akfiny Hasdi Aimon (c), Cici Wulandari (b,c) a) Nuclear Laboratory in Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia *amilaamatullah[at]gmail.com b) Graduate Program in Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia c) Graduate Program in Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia Abstract Small Modular Reactors (SMRs) and micro-reactors show some advantages smaller size and power, accessibility to remote regions, and reduced land requirements. Thorium as an alternative fuel to enhance SMR performance and safety gives some interesting feature such more abundance of resource as well as its superior neutron characteristics and potential for higher fuel efficiency. The study aims to compare uranium and thorium fuel cycles based on small modular integral PWR with thorium base utilization and its comparison with uranium and its combination with additional burnable poison. Initial neutron multiplication factor obtains 1.134 for thorium cycle, 1.129 for uranium fuel and 1.058, uranium-thorium cycles. In addition, Conversion ratio has been analyzed that thorium cycle have a higher CR (Conversion Ratio). with an average CR of 0.765, compared 0.692 for uranium cycles. Furthermore, thorium fuel cycle has significantly lower production of plutonium isotopes compared to other fuel cycles. The study show high power peaking factor (PPF) of 1.923 at the Beginning of Cycle (BOC), requiring further research for mitigation although its still less than 2. Analysis on performance comparisons at different power levels necessitate modifications in terms of geometry, operational duration, or fissile nuclide enrichment has been analyzed for optimization process. This research shows the potential of thorium-based fuel utilization in SMRs, by improving fuel efficiency and reactor safety. Keywords: SMR- Conversion ratio- Neutron multiplication factor- Conversion Ratio- Thorium to Submit This Sample Abstract Share Link | Plain Format | Corresponding Author (Amila Amatullah)

32	Innovative Nuclear Energy Systems	ABS-77
Evaluating 10 MWt Pebble Bed Reactor with Different Nuclear Data Libraries and Monte Carlo-based code OpenMC Kharisma A.D. Rahmany(a), Dwi Irwanto(a,b,c*), Syaiful Bakhri(d), a)Department of physics, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jl. Ganesa No.10, Lb.Siliwangi, Coblong District, Bandung City, West Java 40132,Indonesia b)Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jl. Ganesa No.10, Lb.Siliwangi, Coblong District, Bandung City, West Java 40132,Indonesia c)Nuclear Physics and Biophysics Research Group, Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa No.10, Lb.Siliwangi, Coblong District, Bandung City, West Java 40132, Indonesia d) Research Center for Nuclear Material and Radioactive Waste Technology (PRTBNLR), Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), South Tangerang 15314, Indonesia dirwanto[at]itb.ac.id Abstract HTGR, as a gas-cooled VHTR reactor, is one of the reactors that has been widely researched and developed. It is even considered to have mature technology. Based on the fuel, HTGR has a pebble bed type where this system could perform online refuelling. This study aims to determine the neutronic aspects of the 10 MWt Pebble Bed Reactor, which is based on HTR-10 design, with different nuclear data libraries. The analysis was conducted by reviewing the simulation results of the effective multiplication with JENDFL 3.3, JENDL 5.0, ENDF/B-VII.I, and ENDF/B-VIII.0 nuclear data. The simulation uses the Monte Carlo code, OpenMC, by simulating the entire reactor in the initial criticality state. This study investigates how different nuclear data impact the effective multiplication factor, with a detailed analysis focusing on neutronic parameters. The findings aim to offer a more comprehensive understanding of the development of the Pebble Bed reactor, contributing to advancing nuclear reactor design and safety. Keywords: Pebble Bed Reactor, Nuclear Data Library, Effective Multiplication Factor, Monte Carlo Method Share Link | Plain Format | Corresponding Author (Kharisma Rahmany)

33	Innovative Nuclear Energy Systems	ABS-78
Neutronic and Thermal Hydraulic Analysis of Gas Cooled Fast Reactor in Indonesia - An Overview Kevin Wijaya(a) and Dwi Irwanto(a,b) a) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. b) Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. Abstract Gas-cooled Fast Reactor (GFR) is one of the fourth-generation reactors that is still being developed. The GFR system combines the advantages of a fast spectrum system for long-term sustainability of uranium resources and waste reduction (through fuel reprocessing and long-lived actinide fission) with a high-temperature system (industrial use and high thermal cycle efficiency). Several types of reactors, research methods have been carried out such as the type of output power produced, fuel variations, fuel composition variations, ring geometry variations, and core configuration variations have been carried out. Research has been carried out related to the thermal hydraulic analysis of GFR using various numerical methods. This article discusses the results of several neutronic and thermal hydraulic analyzes from research that has been carried out and discusses things that can be developed from previous GFR research. Keywords: Burn up- fast reactor- GFR- hydraulic thermal- multiplication factor Share Link | Plain Format | Corresponding Author (Kevin Wijaya)

34	Innovative Nuclear Energy Systems	ABS-80
Parametric Analysis of Molten Salt Natural Circulation Loop using Computational Fluid Dynamics Rahmad Dahana Saputra (a), Sidik Permana (a,b*), Duwi Hariyanto (c), Syeilendra Pramuditya (a,b), Syaiful Bakhri (d) (a)Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia (b)Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia (c)Industrial Technology Faculty, Institut Teknologi Sumatera, Jl. Terusan Ryacudu, Lampung, Indonesia (d)Research Center for Nuclear Material and Radioactive Waste Technology (PRTBNLR), Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), South Tangerang 15314, Indonesia Abstract Natural circulation is a form of passive safety in nuclear reactor. Natural circulation is important to prevent another accident like fukushima nuclear disaster that happen because the loss of power to the pump. An experiment on irregular pentagon natural circulation loop has done before using water as its coolant. In this study, the water coolant is replaced with molten salt to analyze its flow parameters. ANSYS Fluent which is a finite volume based computational fluid dynamics software is used to simulate the loop. First the loop heater and cooler temperature is varied to see the respons of the Reynolds number (Re), cold leg temperature (t1), and hot leg temperature (t3) for FLiBe, FLiNaK, and FNaB. For all molten salt, the Re is proportional to heater temperature when the cooler temperature is constant. For constant heater temperature condition, Re does not have a particular trends when the cooler temperature increase. t1 and t3 always increase when the heater or cooler temperature increase. FNaBe is added for second analysis that focuses on comparing Re, t1, and t3 for constant heater condition and constant cooler condition. It is observed that FNaB have the highest Re due to its low viscosity. A correlation on the trends of t1 and t3 to the velocity trends is also observed. Keywords: Natural Circulation, Ansys Fluent, FLiBe, FLiNaK, FNaB, FNaBe Share Link | Plain Format | Corresponding Author (Rahmad Dahana Saputra)

35	Innovative Separation and Fuel Cycles	ABS-2
A Brief Comparison of Indonesian Prospective Nuclear Fuel Cycles Dany Mulyana Center of Nuclear Security Science and Policy Institute, Texas Engineering Experiment Station, Texas A&M University, College Station, TX, USA. Research Center for Nuclear Reactor Technology, Badan Riset dan Inovasi Nasional, Jakarta, Indonesia. Abstract A brief comparison study on fuel cycles of four prospective Indonesian power reactors has been completed. The study oversees 160 MWt NuScale Small Modular Reactor (SMR), 10 MWt Reactor Daya Eksperimental (RDE-10), 40 MWt Pembangkit Listrik dan Uap Panas Industri (PeLUIt-40) and Thorcon^s TMSR-500. Both RDE-10 and PeLUIt-40 were fueled with 17wt% HALEU while the other two were fueled by 4.95wt% LEU, without thorium. The study was based on a set of data generated through Monte Carlo reactor physics simulation using OpenMC. The study found that NuScale SMR is the least 235U consumer by 0.84 g.MW-1 d-1 while the RDE-10 is the most 235U consumer by 0.97 g.MW-1 d-1. The simulation results show that TMSR-500 is the most uranium consumer by 1.41 g.MW-1 d-1 while the uprating of RDE-10 to PeLUIt-40 makes PeLUIt-40 as the least uranium consumer by 1.14 g.MW-1 d-1. As the least uranium consumer, both RDE-10 and PeLUIt-40 provide the highest attainable discharged fuel burnup of extractable energy of about 146 GWd/MTU. Assuming a one batch depletion scheme, NuScale SMR, RDE-10, PeLUIt-40 TMSR-500 has fuel consumption cycles in 2,210 days, 2140 days, 550 days, and 320 days, respectively. Keywords: Fuel Cycle, Indonesian Nuclear Power Plant, NuScale SMR, PeLUIt-40,Reaktor Daya Eksperimental, TMSR-500 Share Link | Plain Format | Corresponding Author (Dany Mulyana)

36	Innovative Transmutation Systems	ABS-12
Obtaining the Optimum TRU Layer Configuration for Transmutation in Molten Salt Reactor Fuel Channel R. Andika Putra Dwijayanto 1, Andang Widi Harto 2, Azizul Khakim 1 1 Research Centre for Nuclear Reactor Technology, Research Organisation for Nuclear Energy, National Research and Innovation Agency, 80 KST BJ Habibie, South Tangerang, Indonesia 15343 2 Department of Nuclear Engineering and Physics Engineering, Faculty of Engineering, Universitas Gadjah Mada, Jl. Grafika No. 2, Yogyakarta, Indonesia 55281 Abstract Incineration of transuranic (TRU) elements can be performed in various types of nuclear reactor. Thermal molten salt reactor (MSR) can be utilised for TRU incineration, typically by blending TRU elements with the carrier salt. This method has issues in low trifluoride solubility and lower TRU burnup than fast spectrum MSR. This study proposes an innovative method of TRU incineration in MSR as a heterogeneous configuration by inserting tubular TRU layer into MSR fuel channel. To obtain optimum configuration, TRU layers comprised of reactor grade plutonium (RGPu) and minor actinide (MA) were inserted in an MSR fuel channel in central, middle, and outer configurations. Neutronic analysis was performed in the MSR fuel channel using MCNP6.2 code, and decay analysis was done using ORIGEN2.1. The analysed parameters include infinite multiplication factor, temperature coefficient of reactivity (TCR), transmutation efficiency (TE), and post-irradiation decay activity. It was observed that TRU layer with a volume fraction of 5% in a narrow fuel channel and outer configuration exhibit the optimum TE and improved the TCR. After 1460 days of irradiation, the TE was found to be 30.98% and lower radioactivity than unirradiated TRU after 400 years of decay. This result is promising as an initial stage prior to whole core calculation. Keywords: SD-TMSR, MCNP6.2, TRU incineration, transmutation efficiency Share Link | Plain Format | Corresponding Author (R Andika Putra Dwijayanto)

37	Material and Process for Innovative Energy Systems	ABS-3
Computed Radiography Method for Non-Destructive Testing to Characterize Cooling Piping in Research Reactor G.A. Siwabessy Jepri Sutanto, Baskan Hanurajie, Iwan Sumirat, Sidik Permana, Asril P.A Mustari &#305- ITB BRIN Abstract A radiographic test is one of the NDT methods used to verify the material (non-destructive testing). It utilizes ionizing radiation, such as X-rays or gamma rays, to create detailed images that reveal hidden flaws, defects, discontinuities, or anomalies. A radiographic test method commonly checks welding connections, fabrication, forging, and casting. Radiography NDT technique is used across diverse sectors, including manufacturing, aerospace, infrastructure, energy, and even research reactors or nuclear power plants (NPP). This study will discuss piping in a research reactor in South Tangerang, i.e., G.A. Siwabessy. Industrial digital radiography substitutes conventional film radiography and is replaced by imaging plate (IP). Parameters such as SNR and CNR and basic spatial resolution (SRb) must be considered and understood. Digital radiography is currently more widely used in addition to the short process and interpretation of results with up-to-date software developments. The linear indication is attached to the result of the interpretation. The linear indication^s orientation, size, and shape recommendations can provide valuable information to the nuclear energy research organization while replacing or substituting the piping. Image interpretation results were obtained to provide recommendations for improvements by standards and regulations. Other NDT methods are needed to compare the results. Keywords: Non-destructive testing, industrial radiograph, x-ray, material, computed radiography Share Link | Plain Format | Corresponding Author (Jepri Sutanto)

38	Material and Process for Innovative Energy Systems	ABS-5
Combined Gamma Irradiation and Sodium Lauryl Sulphate (SLS) Surfactant Methods as Modified Activated Carbon (g-SLS/AC) for Fe2+ Wastewater Adsorbent Alvino Andreas Lumban Tobing, Sherina Massayu Putri, Angelica Isabella Christian, Sidik Permana, Asril Pramutadi, Dhita Ariyanti Nuclear Chemical Engineering, Polytechnic Institute of Nuclear Technology, National Research and Innovation Agency, BRIN. Department of Doctoral Nuclear Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha 10 Street, Bandung, 40132, Indonesia Abstract Heavy metal waste was one of big threat because of its negative impact on human health and environment. Heavy metals such as iron, Fe, was found in industrial wastewater, so it was necessary to reduce thus level or concentration. The use of activated carbon was one of the effective steps to carry out the sorption of Fe ions. Therefore, modification of activated carbon was carried out in this study to improve the quality of Fe2+ ion sorption on activated carbon. The methods of this research was devided on three steps. First stages was preparation of activated carbon form of sieving using a 70 mesh sieve and contacting the activated carbon with SLS surfactant for 24 hours. The second steps was exposured activated carbon modified using various dose gamma irradiation, then drying. The third steps was found out optimum contact time. Optimum contact time varied on 15, 30, 45, and 60 minutes. And the last stages was identification optimum dose irradiation. In this research, irradiation dose gamma used was 0- 10- 30- and 50 kGy. The results show that activated carbon that has been irradiated and contacted with SLS had much better sorption capacity than activated carbon it self. Capacity sorption of time contact 15 minutes was 0,8455 mg/g- 30 minutes was 0,9915 mg/g- 45 minutes was 0,9135 mg/g- and 60 minutes was 0,7305 mg/g. So, the optimum contact time was 30 minutes (0,9915 mg/g). Capacity sorption of dose gamma irradiation on 0 kGy was 0,66095 mg/g- 10 kGy was 0,86365 mg/g- 30 kGy was 0,800911 mg/g- and 50 kGy was 0,85125 mg/g. The conclusion of this research is modified activated carbon used gamma irradiation exposure and surfactant had better capacity sorption, by optimum time contact is 30 minutes (0,9915 mg/g) and 10 kGy (0,86365 mg/g) dose irradiation gamma. This research is expected to contribute to improving water quality. Keywords: gamma irradiation, surfactant, adsorption, Fe2+ wastewater Share Link | Plain Format | Corresponding Author (Dhita Ariyanti)

39	Material and Process for Innovative Energy Systems	ABS-6
Methyl Ester Sulfonate (MES) Surfactant and Gamma Irradiation Modified Activated Carbon (g-MES/AC) as Cu2+ Wastewater Adsorbent Sherina Massayu Putri (1), Alvino Andreas Lumban Tobing (2), Angelica Isabella Christian (3), Sidik Permana (4), Asril Pramutadi (5), Deni Swantomo (6), Dhita Ariyanti (7) Nuclear Chemical Engineering, Politechnic Institute of Nuclear Technology, National Research and Innovation Agency, BRIN Department of Doctoral Nuclear Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha 10 Street, Bandung, 40132, Indonesia Abstract Activated carbon is currently popular in advanced materials research and still continuosly improvement. One of activated carbon applications is on quality decrease problem solving of water, specially on presence of heavy metal waste, Cu2+ that caused by industrial activities. Therefore, the combination of activated carbon with surfactant Methyl Ester Sulfonate (MES) and gamma irradiation method is intended to adsorb Cu2+ cationic. In this research, activated carbon was treated by addition of surfactant Methyl Ester Sulfonate (MES) using various dose gamma irradiation 0- 10- 30- and 50 kGy. The first stages carried out preparation of activated carbon form of sieving using a 70 mesh sieve and contacting the activated carbon with MES surfactant for 24 hours. The second stages was exposured activated carbon modified using various dose gamma irradiation, then drying. The third stages was found out optimum contact time. Optimum contact time varied on 15, 30, 45, and 60 minutes. And the last stages was identification optimum dose irradiation. The results showed that Methyl Ester Sulfonate (MES) surfactant and gamma irradiation modified activated carbon adsorbs Cu2+ wastewater effectively. The adsorption capacity was 1,037 mg/gr on 15 minutes- 0,7075 mg/gr on 30 minutes- 0,762 mg/gr on 45 minutes- and the lower was 0,4725 mg/gr on 60 minutes. Optimum contact time of surfactant and gamma irradiation modified activated carbon to adsorbed Cu2+ was 15 minutes. The adsorption capacity was 0,15 mg/gram on 0 kGy- 0,803 mg/gram on 10 kGy- 0,098 mg/gram on 30 kGy- and 0,01 mg/gram on 50 kGy. So, the optimum dose irradiation of surfactant and gamma irradiation modified activated carbon to adsorbed Cu2+ was 10 kGy. From the result, it can be concluded that optimum condition of Methyl Ester Sulfonate (MES) surfactant and gamma irradiation modified activated carbon to adsorbed Cu2+ was 15 minutes and 10 kGy dose gamma irradiation. Hopefully, in the future, the use of irradiated activated carbon is expected to make clean water management more effective and efficient. Keywords: gamma irradiation, surfactant, adsorption, Cu2+ wastewater Share Link | Plain Format | Corresponding Author (Dhita Ariyanti)

40	Material and Process for Innovative Energy Systems	ABS-16
ANALYSIS THE EFFECT OF GAMMA IRRADIATION DOSE VARIATION CO-60 ON THE RESULTS OF BIODIESEL RANDEMEN WITH A CONTENT OF 100% CRUDE PALM OIL (CPO) THROUGH THE USE OF COOKING OIL Renaldy Bernardo Saragih, Wilada Nafi Royani, Dhita Ariyanti Department of Nuclear Chemical Engineering, Polytechnic Institute of Nuclear Technology, National Research and Innovation Agency, Babarsari Street PO BOX 6101 YKBB Yogyakarta 55281, Indonesia Directorate of Laboratory Management, Research Facilities, and Science and Technology Park, National Research and Innovation Agency, Yogyakarta 55281, Indonesia Abstract Biodiesel is an alternative fuel produced from renewable vegetable or animal sources. This fuel has the ability to replace or mix with conventional diesel, with the aim of reducing exhaust emissions that harm the environment. This study aims to analyze the level of influence of cobalt-60 gamma irradiation on the results of randemen in the manufacture of biodiesel based on cooking oil or Crude Palm Oil. Biodiesel is synthesized from cooking oil through two main processes. First, through esterification using methanol as a solvent in a mole ratio of 1:6, involving an acid catalyst at a concentration of 0.05% by weight of oil. Second, through transesterification with methanol in a mole ratio of 1: 6 with a base catalyst, with a catalyst concentration used of 1% by weight of oil. Characterization of biodiesel from synthesis was carried out using FTIR spectrophotometer analysis technique, which identified the presence of ether and methyl functional groups as characteristic. The results of the analysis showed the presence of C-O bonds at wavenumbers 1020/cm-1040/cm and C-H bonds at wavenumbers 2980.88 / cm. Furthermore, the test was conducted in accordance with the Indonesian national standard (SNI) 7182-2015. The test results showed the results of biodiesel randemen at irradiation doses of 35 Kgy has the largest volume of 108 mL with a density of 831 kg/m3, kinematic viscosity of 5,780 mm2 / s at 40 &#913-C, as well as the flame color reddish blue and slightly smoky, through gamma irradiation, biodiesel production can be increased in terms of production either with or without a mixture of other additional compounds that offer potential as an alternative fuel that is environmentally friendly and performs well. Keywords: biodiesel, cooking oil, fuel, irradiation, gamma Share Link | Plain Format | Corresponding Author (Dhita Ariyanti)

41	Material and Process for Innovative Energy Systems	ABS-21
Comparison of Fission and Non-Fission Methods in Mo-99 Radioisotope Production Dhiya Salma Salsabila (a), Anis Rohanda (b) a. Master Program in Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. b. Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, Building No. 80, BJ Habibie Integrated Science Area, South Tangerang 15310, Indonesia Abstract The utilization of the G.A. Siwabessy Multipurpose Reactor (RSG-GAS) includes the production of radioisotopes. Among the radioisotopes produced by RSG-GAS is Mo-99. The Mo-99 radioisotope is used in the production of Technetium-99m (99mTc), which is used in diagnostic imaging in the medical field. There are two ways to produce the Mo-99 radioisotope. First, with a fission scheme using Low Enriched Uranium (LEU) in the form of UO2. Second, with a non-fission scheme by irradiating MoO3 samples. The aim of this study is to analyze the production results of radioisotopes using the fission scheme compared to the non-fission scheme using the ORIGEN 2.1 program. In this study, an MoO3 sample weighing 8.753 grams is compared to UO2 weighing 8.753 grams with the composition: U-235 1.508 grams, U-238 6.206 grams, and O-16 1.039 grams. Then, variations in the irradiation position in the reactor were made, namely at the Central Irradiation Position (CIP) at locations E7 and D6, as well as the Irradiation Position (IP) at point G7 with reactor power of 5, 15, and 30 MWt. The calculation results show that at CIP-D6 with a power of 30 MWt, the production result of the non-fission scheme is 7.05 Ci. Meanwhile, the production result of the fission scheme under the same conditions is 797.10 Ci. This is due to the better ability of UO2 to capture neutrons needed in nuclear reactions, as well as having more uranium isotopes that easily undergo neutron fission. Therefore, for Mo-99 production, it is recommended to use the fission scheme. Keywords: RSG-GAS- Irradiation- Radioisotope- Mo-99 Share Link | Plain Format | Corresponding Author (Dhiya Salma Salsabila)

42	Material and Process for Innovative Energy Systems	ABS-23
Analysis of Fuel Salt Relocation from the Reactor Vessel to the Drain Tank in Molten Salt Reactor (MSR) with Moving Particle Semi-Implicit (MPS) Method Aziz Satrio, Muhamad Daffa Fawwaz, Yulia Mifftah Huljanah, Asril Pramutadi Andi Mustari Bandung Institute of Technology Abstract The Molten Salt Reactor (MSR) is a fourth-generation reactor whose technology has been proven by MSRE^s operations at ORNL. This reactor has a highly inherent safety system. Then, it has ability to extinguish multiple reactors and is carried out passively. Fuel salt in MSR can flow to drain tank by opening melt valve when temperature exceeds the limit. Understanding relocation of fuel salt is essential to improve reactor safety and performance. In this study, we present a particle-based approach using the Moving Particle Semi-Implicit (MPS) method, which represents fluid as particles that carry physical properties. We analysed the dynamics of several types of fuel salt at MSR. The fuel salt variations tested were LiF-NaF-KF, LiF-BeF2, KCl-MgCl2, and NaNO3-NaNO2-KNO3. In addition, temperature variations of 700K, 850K, and 950K were carried out on LiF-NaF-KF. The simulation uses the 2D geometry of MSRE reactor design. This simulation results show that NaNO3-NaNO2-KNO3 has movement with the greatest speed compared to other fuel salt. At 950K for LiF-NaF-KF has the greatest velocity compared to other variations. These results can be taken into account in selection of right saline fuel mixture and the understanding of its flow characteristics for safe and effective MSR operation, which is crucial for improving the safety of nuclear reactors. Keywords: Drain Tank, Fuel Salt, MPS method, MSR Share Link | Plain Format | Corresponding Author (Yulia Mifftah Huljanah)

43	Material and Process for Innovative Energy Systems	ABS-48
Geometrical and Design Analysis of Betavoltaic Nuclear Battery using Monte Carlo Simulation Khairul Basar, Rheza Pahlevi Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung Abstract Betavoltaic nuclear batteries are one of the energy devices that have future use prospects. The operational efficiency of betavoltaic nuclear batteries is still quite small, which is caused by the geometrical design, or the semiconductor materials used. In this study, Monte Carlo simulation was used to determine the deposition energy of beta particles in semiconductor materials in betavoltaic nuclear battery systems. The performance of betavoltaic nuclear batteries was determined using the current-voltage characteristic curve on the semiconductor and beta particle parameters obtained from Monte Carlo simulations. Keywords: betavoltaic, nuclear battery, Monte Carlo simulation Share Link | Plain Format | Corresponding Author (Khairul Basar)

44	Material and Process for Innovative Energy Systems	ABS-49
Study Aging component reactor of Al6061T as lining reactor TRIGA 2000 in demin water contaminated By NaCl evaluation by outpile corrosion testing Djoko Hadi Prajitno, Irawan Sugoro , Wahyu Ramdhani and Dani Gustaman Syarif National Research and innovation Agency Abstract In order to increase safety, nuclear component of the reactor have got good performance on normal operational conditions. One of nuclear component of the reactor is lining tank reactor. In this study, the aging component reactor of Aluminum alloy Al6061 as lining tank reactor Triga 2000 in demin water has been evaluation by using electrochemical methods. The electrochemical testing were performed in demin water with and without the addition NaCl as contaminant at ambient temperature. The electrochemical technique used in this study were open circuit potential (OCP), Tafel polarization and electrochemical impedance spectroscopy (EIS). The results of open circuit potential tests showed that the corrosion potential (Ecorr) values of Al6061 T alloy are slightly different between in demin water with and without the addition NaCl. The corrosion potential Al6061 T alloy are higher in demin water without NaCl compare with NaCl addition. The are show that the interaction between demin water and the aluminum alloy have no significant effect on to its corrosion behavior. However, Tafel polarisation of aluminum alloy Al6061 T alloy in demin water with the absence of NaCl showed the lowest corrosion than aluminum alloy at condition in demin water with NaCl. Eis study shows that aluminum alloy Al6061 T alloy in demin water with the absence of NaCl showed the higher impedance than aluminum alloy in demin water with NaCl. The higher impedance indicated that the passive film of Al2O3 formed on the surface of aluminum alloy. In terms of corrosion rate values, aluminum alloy Al6061T showed the best corrosion resistance in demin water without NaCl compare Al 6061 T in demin water with NaCl. Keywords: Please Just Try to SubmiAl6061 T ,corrosion, demin water, NaClt This Sample Abstract Share Link | Plain Format | Corresponding Author (Djoko Hadi Prajitno)

45	Material and Process for Innovative Energy Systems	ABS-59
Microstructure Evaluation and Rietveld Analysis of Fe-Cr-Al based ODS Alloy using Arc Melt Casting Method Muhammad Angga Saputra(a)- Andriansyah(b)- Diene Noor Haerani(b)- Rohmad Salam(c)- Muhammad Fakhrudin(d)- Syahfandi Ahda(b)- Bernardus Bandriyana(b)- Tetty Kemala(a)- Nanda Shabrina(b)* (a) Department of Chemistry, Faculty of Mathematics and Natural Sciences (FMIPA), IPB University (b) Research Centre for Rector Nuclear Technology (PRTRN), Research and Organization for Nuclear Energy (ORTN), National Research and Innovation Agency of Indonesia (BRIN) (c) Directorat for nuclear facility management (DPFK), National Research and Innovation Agency of Indonesia (BRIN) (d) Research Centre for Advanced Material (PRMM), Research Organization for Nanotechnology and Material (ORNM), National Research and Innovation Agency of Indonesia (BRIN) *nanda.shabrina[at]brin.go.id Abstract Fe-Cr-Al-based ODS (Oxide Dispersion Strengthened) alloy emerging as a promising candidate for fuel cladding applications. The ODS steel was fabricated by Arc Melt Casting as alternative method beside the common sintering method. The effect of yttria dispersion (0.5% and 3%) in Fe-Cr-Al matrix was investigated. Crystal structure, microstructure, element analysis and hardness were carried out using XRD (X-Ray Diffraction), Optical Microscopy, SEM-EDS (Scanning Electron Microscopy) and Vickers Hardness Test. Results shows by increasing 3% Y2O3, grain size area decrease 50% following by hardness number increase 23%. Crystal structure identification was analysis by Rietveld analysis using PAN-Analytical Highscore Software. In 0.5% Y2O3 addition, Alumunium-Iron found to be dominant beside Chromium Iron and Chromium. However, the analysis from EDS showed agglomeration of Al and Y2O3. Keywords: ODS- Fe-Cr-Al- Y2O3- alloy- Arc melt casting Share Link | Plain Format | Corresponding Author (Nanda Shabrina)

46	Nuclear Data	ABS-1
Evaluation of nuclear reaction cross sections for optimization of production of the emerging diagnostic radionuclide 123I via proton and deuteron-particle induced transmutations Lutfi Aditya Hasnowo1,2 1 Polytechnic Institute of Nuclear Technology, National Research and Innovation Agency, Yogyakarta, 55281, Indonesia 2 School of Nuclear Science and Engineering, Tomsk Polytechnic University, Tomsk, 634050, Russia Abstract Proton and deuteron-particle induced reactions on 122,123,124Te targets were evaluated for the production of 123I. The literature data were compared with nuclear model calculations using the codes TALYS-2.0 and EMPIRE-3.2.3 (Malta). The statistically evaluated excitation functions were generated- therefrom the integral yields of the 123I were calculated. The amounts of the radioactive impurities were assessed. Keywords: 122Te(d,n)123I, 123Te(d,2n)123I, 123Te(p,n)123I, 124Te(p,2n)123, Nuclear model calculations, Nuclear data evaluation, Thick target yields Share Link | Plain Format | Corresponding Author (Lutfi Aditya Hasnowo)

47	Nuclear Data	ABS-38
Exploring of of Macro-Micro Minerals in Different Rice Species at Pandenglang Utilizing Neutron Activation Analysis Method Sari Hasnah Dewi, Th. Rina Mulyaningsih, Saeful Yusuf, Alfian, Istanto, Siti Suprapti, Firda Amalia, Fatmawati and Ferly Hermana Organization Research for Nuclear Energy - National Research and Innovation Agency Abstract Exploring of macro-micro minerals in different rice species produced in Pandeglang had been done using neutron activation analysis. Neutron activation analysis is a method using neutron particle which produced in nuclear reactor then be shoot into sample atom and then the sample is active and scattered Gamma Ray. Furthermore gamma ray detected by gamma detector. Kind of sample of rice have been studied were planted in city around Pandeglang city that are white rice Sobang, IR 12, IR 64, Cimanuk, Ciherang, Red Rice Huma, Red Rice, Black Sticky Rice Cimanuk, Sticky Rice Serang and Sticky Rice Karang Tanjung. Analysis showed samples contained macro mineral more than 100 mg/Kg, such as K, Ca, Mg, Na, and Cl. Contain of micro minerals in sample were in between 0.2 < x >30 mg/Kg : Zn, Cr, Co Br, Rb and Mn. Surprisingly all of those kind of rice have contained other mineral that not include in macro or micro minerals category such as Al, Cs and La in massive amount more than 5 mg/Kg. These minerals are literally not needed in human consumption and metabolism, it must be take into account into serious attention because of its toxicity. Evaluation of these elements is compared to sufficiency value of daily requirement RDA (Recommended Daily Acceptable). Among the concerns is the high content of many kind of metals in colored rice , especially Black Sticky Rice Cimanuk meanwhile for white rice species relatively contain low concentration of either micro-macro mineral and toxic element. In this study was discussed potential hazards for human while its deficiencies or excessive intake Keywords: rice, macro-micro minerals, NAA Share Link | Plain Format | Corresponding Author (Sari Hasnah Dewi)

48	Nuclear Education and Social Aspects	ABS-74
Development Module of Natural Nuclear Radiation Analysis for High School Level Intan Fauziyyah, Sidik Permana, Zulfahmi, Ismail Humolungo, Adi rachmansyah, Dwi Irwanto, Nurhasan Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesha 10, Bandung, 40132, Indonesia Abstract Improving scientific literacy among high school students in Indonesia remains a challenge, prompting initiatives to enhance this skill. This research aims to develop a nuclear radiation practicum module for high school students and analyze environmental radiation dose rates. The study involves comparing radiation measurements using various tools and evaluating the impact of environmental background conditions on these measurements. It utilizes Cs-137 (Cesium 137) sources at different distances and with different shielding materials. The ADDIE approach-Analysis, Design, Development, Implementation, and Evaluation-guides the module development process.The study is limited to the development stage, focusing on creating models for the measurement process, mapping, and comparing tools and practicum materials. Radiation dose rate measurements were conducted at the Bandung Institute of Technology at 10 points with instrument heights of 0 cm and 100 cm from the ground, using GMC 500+ and Pocket Geiger tools. Results show that the GMC 500+ tool measured an average radiation dose rate of 0.1365 \pm 0.0160 \mu Sv/hour at 0 cm height and 0.1207\pm0.0163 \muSv/hour at 100 cm height. The Pocket Geiger tool recorded an average dose rate of 0.052\pm0.0154 \mu Sv/hour at 0 cm and 0.043\pm0.0176 \mu Sv/hour at 100 cm height. Measurements on different background surfaces (granite floors, paving blocks, soil, concrete, grass, and asphalt). Measurements using the Cs-137 source determine the effect of distance on the radiation source and shielding material, with distance variations from 0 cm to 100 cm and the use of shielding material stored between the source and the detectorThe final product is a validated nuclear radiation practicum module, which received positive feedback from material and media experts, peers, and high school physics teachers, indicating its readiness for implementation. Keywords: Geiger Muller- Pocket Geiger- Environmental Radiation- Cesium 137, Practical Module Share Link | Plain Format | Corresponding Author (Intan Fauziyyah)

49	Nuclear Nonproliferation Issues	ABS-4
Intrinsic Nuclear Security Attribute Analysis for Physical Protection System Development of Reaktor Daya Eksperimental Dany Mulyana Center of Nuclear Security Science and Policy Institute, Texas Engineering Experiment Station, Texas A&M University, College Station, TX, USA. Abstract A study on material theft analysis of fresh and spent fuels of Reaktor Daya Eksperimental (RDE) has been completed. This study covers an integrated analysis of nuclear safety, security, and safeguards. This study assumed that each pebble fuel contains about 7,223 tri-structural isotropic (TRISO) fuel kernel microspheres depleted up to 90 GWd/MTU. Both the TRISO and the graphite matrix of the fuel pebble was designed to contain fission products during and after irradiation. A set of Monte Carlo n-Particle (MCNP) neutronic simulations on the RDE^s fresh and spent fuel pebbles were completed to investigate the potential radiation exposure to the personnel, that can be a security barrier for RDE system. Taking only gamma ray into considerations, this study found that the fresh fuel pebble containing 5 g of UO2 enriched to 17 wt% of U-235 exposes radiation with a dose rate of only 0.38 microrem/h at a distance of 1 m away from the pebble. At the same distance, a 90 GWd/MTU spent fuel pebble cooled for one year resulted dose rate of about 1.0 rem/h. This study also found that a lead sphere shielding (density of 11.34 g/cm3) with a thickness of approx. 12.5 cm must cover the spent fuel pebble to reduce dose rate to 0.328 microrem/h. Keywords: Nuclear Security, Physical Protection System, Pebble Bed Reactor, Reaktor Daya Eksperimental Share Link | Plain Format | Corresponding Author (Dany Mulyana)

50	Nuclear Nonproliferation Issues	ABS-19
Burnup Performance of modified CANDLE shuffling in axial direction on gas cooled fast reactors with UN-Th fuels Jean Pierre Ndayiragije, Zaki Su^ud, Abdul Waris, Dwi Irwanto Institut Teknologi Bandung Abstract The modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) strategy shuffling in an axial direction has been successfully applied to gas cooled fast reactor. This paper investigated the utilization of natural uranium, enriched nitride and Thorium (238U 15N and 232Th) as fuel on 3000MWt reactor power and refuelling every 10 years of burnup cooled by helium gas. The reactor core is partitioned into ten equal-volume regions in the axial direction. Initially, a fuel input (238U 15N and 232Th) is placed in region 1. After ten years of burnup, a fuel in region 1 has been moved to region 2, then to region 3, and so on until the fuel in region 9 was moved to region 10. A fuel from region 10 has been got out. The neutronic computations were performed in two different ways utilizing the SRAC 2006 code and JENDL4.0 as a nuclear data library. Firstly, PIJ has been used for fuel cell calculations and secondly CITATION for reactor core calculation. The results show that the effective multiplication factor is greater than one, this indicates that the reactor is capable to operate through a burn-up period by employing 238U 15N and 232Th as fuel cycle input and the burnup level at the end life is about 30.29% HM. Keywords: Thorium, Modified CANDLE, SRAC, Effective multiplication factor, burnup Share Link | Plain Format | Corresponding Author (Jean Pierre Ndayiragije)

51	Nuclear Nonproliferation Issues	ABS-33
Development of environmental and radiation monitoring system based on Private LoRaWAN Network I Putu Susila (a*), Gina Kusuma (a), Dian Fitri Atmoko (a), Arya Adhyaksa Waskita (b), Gede Arna Jude Saskara (c) a) Research Center for Nuclear Beam Analysis Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency KST BJ. Habibie, Tangerang Selatan 15310, Indonesia *) i.putu.susila[at]brin.go.id b) Research Center for Data and Information Sciences, Research Organization for Electronics and Informatics, National Research and Innovation Agency KST Samaun Samadikun, Jalan Sangkuriang, Bandung 40135, Indonesia c) Department of Information System, Faculty of Engineering and Vocational, Universitas Pendidikan Ganesha Jalan Udayana 11, Singaraja 81116, Indonesia Abstract A scalable private LoRaWAN-based monitoring system has been developed for indoor and outdoor radiation monitoring around nuclear facility. The system consists of open hardware-based radiation detection devices, indoor and outdoor LoRaWAN gateways, an open-source LoRaWAN network server and database server for storing measurement data. In addition, a commercial air quality monitoring device has been integrated to the proposed system to ensure its scalability. The aims of the system are to evaluate and validate the feasibility of low-cost open hardware based radiation measurement devices, development of private LoRaWAN based infrastructure for the monitoring data pipeline, and provide testbed of wireless-based security for radiation detection system. The system has been deployed and tested to measure the coverage of LoRaWAN network, and performance measurement of real-time radiation and environmental data collection for indoor and outdoor measurement. The result demonstrates that the maximum coverage distance between gateway and end-device is ~600 m depend on terrain, and the continuous transmission of real-time measurement data i.e. radiation dose rate, environmental and air quality data can be done within a few seconds interval. In the future, security assessment of the proposed system need to be performed to ensure the integrity and reliability of the system when deployed as a measurement system for protecting environment and people from the risk of radiation exposure. Keywords: Radiation safety and security- Radiation monitoring- Environmental monitoring- LoRaWAN Share Link | Plain Format | Corresponding Author (I Putu Susila)

52	Nuclear Nonproliferation Issues	ABS-75
Assessment of Material Attractiveness in Light Water-Based Reactors: Evaluating Plutonium Isotope Composition and Burnup Impact on Proliferation Risk Fungky Iqlima Nasyidiah(a), Sidik Permana(a,b*), Dwi Irwanto(a,b), Akfiny Hasdi Aimon(c), Cici Wulandari(a,b) a)Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia b)Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia (c) Physics and Technology of Advanced Materials Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia *psidik[at]itb.ac.id Abstract The assessment of nuclear safety and the potential risks associated with proliferation is a critical area of research in the context of nuclear energy utilization. This study focuses on evaluating the Material Attractiveness (ATTR) in several light water-based reactor designs, specifically NuScale, ESBWR, BWRX-300, and PWR, both during reactor operation and post-operation. The analysis employed Origen 2.2 for modeling radioactive decay and MCNP4C for assessing material attractiveness. The findings reveal an increase in the production of plutonium isotopes, with the exception of Pu-239, which decreases as a result of the fission. Initially, the Pu-240 composition across all four reactors is classified as Super-grade plutonium. However, with increased burnup, the Pu-240 composition transitions to a reactor-grade plutonium level. At the initial of irradiation, the ATTR values were determined to be 0.19 for the ESBWR, 0.20 for the PWR, 0.16 for the BWRX-300, and 0.21 for NuScale, categorizing them within the weapon-grade range. By the end of reactor operation, these values had significantly decreased, with ESBWR at 0.0125, PWR at 0.0148, BWRX-300 at 0.0111, and NuScale at 0.0133, placing them in the un-usable grade category. This marked reduction in ATTR values, correlated with increased burnup, indicates an effective decrease from weapon-grade to un-usable grade by the end of the reactor^s operational period. This trend is primarily influenced by the increased production of isotopes Pu-238, Pu-240, Pu-242, and the corresponding decrease in Pu-239, the primary fissile material. Keywords: Non-proliferation- ATTR- Bare Critical Mass- Decay Heat- Spontaneous Fission Neutron- Neutron Prompt Life Share Link | Plain Format | Corresponding Author (Cici Wulandari)

53	Radiation Physics	ABS-9
Analysis of Radionuclide Content in Soil Samples Using the X-Ray Fluorescence Spectroscopy (XRF) Method Ismail Humolungo (a*), Sidik permana (a), Adi R. A Abdullah (a), Zulfahmi (a), Haryo Seno (b), Alan Maulana (b) a) Doctoral Program in Nuclear Engineering Department, Bandung Institute of Technology, Jl. Ganesha No. 10, Bandung, Indonesia. b) Indonesias National Research and Inovation Agency. Jl. Tamansari No. 71, Bandung, Indonesia. *ihumolungo41[at]gmail.com Abstract The high average value of natural gamma radiation dose rate in Mamuju Regency indicates that the uniqueness is related to the high concentration of radioactive elements of Uranium, Thorium, and Potassium deposited in the rock. In this study, XRF tests were carried out with EDXRF equipment on 5 soil samples from Mamuju Regency. Measurements were made on powder and pellet specimens. The results showed that the average concentration of radioactive elements uranium (U) was 712.17 ppm and 3581.45 for Thorium (Th), while for K was 9127.75 ppm. The thorium element appears to be more dominant with the Th/U ratio reaching 3.99 which is higher than the average Th/U ratio in the Earth^s crust of 3.70. A positive correlation between U and Th was obtained with a correlation coefficient value of 0.60, and a positive correlation was also shown by the relationship between U and Th, indicating the enrichment of U along with the enrichment of Th. Analysis of XRF data and dose rate data shows that there are two locations that show a very close relationship between the dose rate and the concentration of radioactive elements in the sample, while in other samples the influence of land use into gardens or highway strips affects the measured dose rate. Keywords: XRF, Natural Radiation, Mamuju, Dose Rate Share Link | Plain Format | Corresponding Author (Ismail Humolungo)

54	Radiation Physics	ABS-13
Comparative Study of Natural Background Radiation Between Mamuju Regency, West Sulawesi, and West Java Zulfahmi1, Sidik Permana1.2, Adi R.A.Abdullah1, Imam Ghazali Y2, Ismail Humolongo1, Haryo Seno3 1Doctoral Program in Nuclear Engineering, Bandung Institute of Technology, Jl. Ganesha No. 10, Bandung, Indonesia. 2Master Program in Physic, Bandung Institute of Technology, Jl. Ganesha No. 10, Bandung, Indonesia. 3Indonesia^s National Research and Inovation Agency, Jl. Tamansari No. 71, Bandung, Indonesia. Abstract Some areas have unique radiation levels, caused by the high density of buildings or deposits of radioactive minerals. Mamuju district is the area with High Natural Background Radiation (HNBR) despite the low density of the population, while West Java possessing low-level radiation is the area with the largest population in Indonesia. Measurements of dose levels in these two areas were carried out for two weeks covering nine villages in Mamuju and nine districts in West Java. The method used is random sampling to collect data from one meter on the surface, while the tool used was the Ludlum model 19 series 8 Analog Survey meter. The results of radiation measurements were 2.208 mSv/year from 180 data in the Mamuju district and 0.3086 mSv/year in 103 data in West Java. The big difference between the two is due to the presence of radioactive element-carrying minerals in the Mamuju district, while in West Java areas there are lack of radioactive elements mineralization. Keywords: radiation, scintillator, dose rate, Mamuju, west Java, HNBR Share Link | Plain Format | Corresponding Author (Zulfahmi Zulfahmi)

55	Radiation Physics	ABS-39
Characterization of neutron spectrum on the TRIGA 2000 reactor core using a passive single-moderator neutron spectrometer Rasito Tursinah, Sidik Permana, Zaki Su^ud, Alan Maulana, Iso Suwarso, Prasetyo Basuki, Teguh Subekti, Tri C. Laksono, Putu Sukmabuana Nuclear Science and Engineering Department, Bandung Institute of Technology, Jl. Ganeca 10, Bandung, Indonesia Research Organization for Nuclear Energy, National Research and Innovation Agency, B.J. Habibie Science and Technology Park, Setu, Tangerang Selatan, Banten, Indonesia Directorate of Nuclear Energy Facilities Management, National Research and Innovation Agency, Jl. Tamansari 71, Bandung, Indonesia Abstract The neutron spectrum in the reactor core has a shape with a combination of thermal, epithermal and fast neutrons. Neutron spectrum measurements in the reactor core generally use the multi-foil technique. However, measurements using this technique are quite difficult, expensive to manufacture, but provide less accurate results. The SCNS-reactor has been developed which is a single-moderator neutron spectrometer based on a passive neutron detector for measuring the neutron spectrum in the reactor core. Neutron spectrum measurements have been carried out on the surface of the TRIGA 2000 reactor core using a SCNS-reactor operated at 100 W for 30 minutes. The results of these neutron spectrum measurements were compared with simulations using MCNPX with ENDF7 nuclear data for neutron interactions with fuel and reactor core components, as well as S(&#945-,&#946-) for neutron interactions with the water moderator. Keywords: Neutron spectrum- reactor core- TRIGA 2000- SCNS-reactor Share Link | Plain Format | Corresponding Author (Rasito Tursinah)

56	Radiation Physics	ABS-46
Utilization of Nuclear-Based Analytical Techniques for Characterization and Preservation Studies of Cultural Heritage in Indonesia: Current Progress, Challenges, and Future Outlook Moh Mualliful Ilmi Research Centre for Archaeometry, National Research and Innovation Agency of Indonesia (BRIN) Abstract Indonesia boasts a rich tapestry of cultural heritage artifacts spanning from the prehistoric to the modern colonial and post-colonial eras, offering invaluable insights into ancient artistic techniques, materials, and historical periods. Our research group, in collaboration with fellow scholars, has embarked on pioneering efforts to unravel the material complexities of Indonesian cultural artifacts. Leveraging a diverse array of nuclear-based analytical tools such as X-ray fluorescence (XRF), synchrotron X-ray diffraction (SR-XRD), and X-ray absorption near-edge structure spectroscopy (XANES), alongside complementary laboratory techniques including Raman spectroscopy and microscopic observations using optical and scanning electron microscopy with energy dispersive X-ray spectroscopy (SEM-EDS), we have delved into the physicochemical properties of these artifacts. This integrated approach has yielded significant findings on optical characteristics, elemental compositions, mineralogical structures, electronic properties, microstructures, and elemental distributions at microscopic scales across artifact surfaces. Beyond material analysis, our research addresses the degradation processes affecting various cultural heritage items, from rock art and pottery to bones and other materials. Moving forward, our aim is to expand these investigations to comprehensively analyze deterioration mechanisms and devise effective preservation strategies for these irreplaceable cultural treasures. While our efforts have been fruitful, challenges remain, particularly concerning the limitations of portable analytical instruments that hinder on-site non-destructive analysis and micro-sampling. Future directions include the exploration of advanced imaging techniques such as hyperspectral imaging and reconstruction methodologies, which are pivotal for advancing our understanding and preservation efforts. Keywords: Nuclear techniques, Physicochemical, Multianalytical, Characterization, Preservation, Share Link | Plain Format | Corresponding Author (Moh. Mualliful Ilmi)

57	Radiation Physics	ABS-60
Low-Cost Sensor Deployment on a Public Minibus in Fukushima Prefecture Rakotovao Lovanantenaina Omega (a*) , Yo Ishigaki (b) ,Sidik Permana (a),Yoshinori Matsumoto (c),Kayoko Yamamoto (d),Katsumi Shozugawa (e) and Mayumi Hori (f) (a) Graduate Program in Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung City 40132, Indonesia (b) Research Center for Realizing Sustainable Societies, The University of Electro-Communications, Tokyo 182-8585, Japan (c) Department of Applied Physics and Physico-Informatics, Faculty of Science and Technology, Keio University, Yokohama 223-8522, Japan (d) Graduate School of Information Science and Technology Department of Informatics, The University of Electro-Communications, Tokyo 182-8585, Japan (e) Graduate School of Arts and Sciences, The University of Tokyo, Tokyo 153-8902, Japan (f) College of Arts and Sciences, The University of Tokyo, Tokyo 153-8902, Japan Abstract This study examines the implementation of an affordable radiation monitoring system on a public minibus in Fukushima Prefecture. The purpose is to monitor and analyze the levels of radiation in the surrounding area. The study utilizes the Pocket Geiger (POKEGA) sensor, which is combined with a microcontroller and telecommunication system. This integration allows for the gathering and viewing of data in real-time. A monitoring system was implemented on a minibus route between Okuma and Tomioka, which were significantly impacted by the Fukushima Daiichi nuclear accident. The study utilized the Quantum Geographic Information System (QGIS) to calculate the average radiation dose rates over a 1 x 1 km grid. It found a notable decrease in radiation levels from 2022 to 2023. The ecological half-life, determined using land use classification by the Advanced Land Observation Satellite, demonstrated fast radiation decay across different types of land, highlighting the influence of environmental and decontamination parameters. This method not only improves the involvement of the community in monitoring radiation, but also offers information about the success of decontamination efforts and the changing levels of radiation in Fukushima Prefecture. *Abstract adapted from the research article : Low-Cost Sensor Deployment on a Public Minibus in Fukushima Prefecture- by Rakotovao Lovanantenaina Omega ,Yo Ishigaki, Sidik Permana, Yoshinori Matsumoto, Kayoko Yamamoto, Katsumi Shozugawa, and Mayumi Hori, published in MDPI Sensors Keywords: public bus, POKEGA, IoT, ambient dose, half-life Share Link | Plain Format | Corresponding Author (Lovanantenaina Omega Rakotovao)

58	Radiation Physics	ABS-61
Radiation Risks and Assessment Challenges in Industrial and Small-Scale Tin Mining: A Study from Bangka Rakotovao Lovanantenaina Omega (a), Zulfahami (a), Imam Ghazali Yasmint (b), Adi R.A. Abdullah (a), Ismail Humolungo (a), Sidik Permana (a) (a) Graduate Program in Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung City 40132, Indonesia (b) Joint Doctoral Program for Sustainability Research, Graduate School of Informatics and Engineering, The University of Electro-Communications Tokyo, Tokyo 182-8585, Japan Abstract The research conducted in Bangka, Indonesia examines the environmental and health consequences of Naturally Occurring Radioactive Materials (NORM) and Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM) in the area where tin mining processing takes place. The study carried out in March 2023 specifically targets three main locations, which are Mentok, Pangkalpinang, and Sungaliat. Mentok faced enhanced surveillance as a result of its substantial tin mining operations. The quantitative results indicate that the level of ambient radiation exposure from industrial mining in Bangka is considerably greater than that from small-scale mining, with an average disparity of around 2.02 microsieverts per hour. The average annual exposure of workers in some areas was determined to be higher than the global average occupational exposure level of 5 millisievert per year, namely, in tailing. Public places, such as offices are recorded as low exposure. These findings suggest a risk of health hazards caused by radiation exposure, require rigorous safety measures and regulatory supervision for the concerned area. The study offers insights into the environmental radiation levels in the tin mining regions of Bangka, highlighting the necessity of implementing protective measures to ensure the safety of workers and the general public against potential radiation exposure. Keywords: NORM, TENORM, radiation exposure, Bangka Share Link | Plain Format | Corresponding Author (Lovanantenaina Omega Rakotovao)

59	Radiation Physics	ABS-62
Correction Factors Evaluation in Calculation of Saturation Activity for TRIGA 2000 In-Core Neutron Flux Measurements Fahma Roswita(1,3*), Haryo Seno(2), Dikdik Sidik Purnama(2), Nina Widiawati(1), Zaki Su^ud(3), and Nuri Trianti(1) 1) Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency *fahma.roswita[at]brin.go.id 2) Research Center for Nuclear Beam Analysis Technology, Organization for Nuclear Energy, National Research and Innovation Agency 3) Nuclear Science and Engineering Study Program, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Abstract Neutron flux measurement for research reactor in-core is usually using the neutron activation method. The sample in this study used an Au-197 sample. The neutron flux can be obtained by calculating the saturation activity received by the irradiated sample. In the calculation of saturation activity, there are correction factors that can affect the results of the conversion of saturation activity into neutron flux. The calculation of correction factors is carried out analytically by entering variables such as counting efficiency and calculation of uncertainty variables. From this calculation, it can be seen how much influence each component and variable of the correction factor for the saturation activity calculation on the results of neutron flux measurements in the TRIGA 2000 reactor core. Keywords: neutron flux, saturation activity, correction factors, calculation, reactor Share Link | Plain Format | Corresponding Author (Fahma Roswita)

60	Radiation Physics	ABS-67
Early Warning System for Radiological Emergency Preparedness of TRIGA 2000 Reactor Site Santiko Tri Sulaksono (a,b*), Zahra Fauhan Salsabila (c), Danang Supriyanto (d), Haryo Seno (e), Dani Muliawan (d), Gallant Tsany Abdillah (d), Satrio Aris Setiawan (d) a) Department of Doctoral Nuclear Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha 10 Street, Bandung, 40132, Indonesia b) Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia *sant012[at]brin.go.id c) Department of Electrical Engineering, Faculty of Science and Technology, Universitas Muhammadiyah Bandung, Soekarno-Hatta 752 Street, Bandung, 40614, Indonesia d) National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia e) Research Center for Nuclear Beam Analysis Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia Abstract The National Research and Innovation Agency (BRIN) has a nuclear facility, one of which is the TRIGA 2000 research reactor. Radiation protection along with the emergency preparedness and response programs are essentials to be carried out in the nuclear reactor and its surrounding area. These programs require several types of devices to detect and measure radiation. The aim of this study is to obtain the effectiveness of the early warning system that has been developed for radiological emergency preparedness in the TRIGA 2000 reactor site. A Geiger Muller (GM) counter based on the ATMega 328 microcontroller was created for dose rate measurement and an ESP8266 module to send the results to the server. Javascript programming was created to send notifications to smartphones when the measured dose rate exceeds a certain dose rate. Comparison of the dose rate from the GM counter measurement and the data stored on the server showed a small error. This early warning system can also provide notifications to smartphones when the measured dose rate exceeds the threshold. Based on the test results, it can be concluded that the early warning system performed functions effectively to support the radiation protection program in the TRIGA 2000 reactor site. Keywords: radiation protection, early warning system, radiological emergency, Geiger Muller counter Share Link | Plain Format | Corresponding Author (Santiko Tri Sulaksono)

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